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Conference Papers Year : 2019

Identification of fracture parameters for irradiated nuclear fuel

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Abstract

Introduction: This paper deals with the fracture characteristics of UO$_2$ fuel. At low temperature (<900 °C) this ceramic has a brittle behavior in traction. The knowledge of fracture parameters for irradiated UO$_2$ fuel is useful to model and understand the behavior of the fuel in reactor. The crack model used in numerical simulations of the fuel rod behavior is based on two main parameters: the critical stress and the fracture toughness. The aim of this article is to analyze fracture tests to identify these two parameters at different scales. To obtain the rupture parameters, we use two kinds of sample (smooth and notched samples) with three sizes: large (28x4x4 mm$^3$), small (10x1.5x1.5 mm$^3$) and micrometric (13x4x3 $\mu$µm$^3$). Fracture toughness is a local parameter that can be measured using notched samples, so it depends weakly on the sample size. Nevertheless, critical stress seems to depend on the size of the sample. In this article, this problem is discussed using analytical and numerical approaches.
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Dates and versions

cea-02614140 , version 1 (20-05-2020)

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  • HAL Id : cea-02614140 , version 1

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J.-M. Gatt, R. Henry, I. Zacharie-Aubrun, C. Langlois, S. Meille. Identification of fracture parameters for irradiated nuclear fuel. SMIRT25 - 25th Conference on Structural Mechanics in Reactor Technology, Aug 2019, Charlotte, United States. ⟨cea-02614140⟩
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